C<Subscript>60</Subscript>-Fullerene Hyper-Velocity High-Density Plasma Jets for MIF and Disruption Mitigation

We present an innovative idea to use hyper-velocity (>30 km/s) high-density (>10¹⁷ cm⁻³) plasma jets of D-T/H and C₆₀-fullerene for magneto-inertial fusion (MIF), high energy density laboratory plasma (HEDLP), and disruption mitigation in magnetic fusion plasma devices. The mass (~1–2 g) of sublimated C₆₀ and hydrogen (or D-T fuel) produced in a pulsed power source is ionized and accelerated as a plasma slug in a coaxial plasma accelerator. For MIF/HEDLP we propose to create a magnetized plasma target by injecting two high-Mach number high-density jets with fuel (D-T) and liner (C₆₀/C) structure along the axis of a pulsed magnetic mirror. The magnetized target fusion (MTF) plasma created by head-on collision and stagnation of jets is compressed radially by a metallic liner (Z-pinch) and axially by the C₆₀/C liner. For disruption mitigation, the C₆₀ plasma jets were shown to be able to provide the required impurity mass (J Fusion Energy 27:6, 2008).     

Global Energy Confinement Scaling Predictions for Tandem Mirrors

For an MHD stable system, we investigate the interplay between drift wave (ETG and gyro-Bohm) radial transport and axial losses in the GAMMA-10 experimental facility and the proposed kinetically stabilized tandem mirror (KSTM) fusion reactor. Numerical coefficients in the models are taken to be consistent with tokamak and stellarator databases. The trade off between radial losses and the Pastukhov end losses is examined. We propose the use of a genetic algorithm to optimize the fusion power amplification Q = P _fusion/P _injected as a function of the key system parameters.     

New Evaluated Semi-Empirical Formula Using Optical Model for 14–15 MeV (<Emphasis Type="Italic">n</Emphasis>, <Emphasis Type="Italic">t</Emphasis>) Reaction Cross Sections

In the next century the world will face the need for new energy sources. Nuclear fusion can be one of the most attractive sources of energy from the viewpoint of safety and minimal environmental impact. Fusion will not produce CO₂ or SO₂ and thus will not contribute to global warming or acid rain. Achieving acceptable performance for a fusion power system in the areas of economics, safety and environmental acceptability, is critically dependent on performance of the blanket and diverter systems which are the primary heat recovery, plasma purification, and tritium breeding systems. Tritium self-sufficiency must be maintained for a commercial power plant. The hybrid reactor is a combination of the fusion and fission processes. For self-sustaining (D-T) fusion driver tritium breeding ratio should be greater than 1.05. So working out the systematics of (n, t) reaction cross-sections are of great importance for the definition of the excitation function character for the given reaction taking place on various nuclei at energies up to 20 MeV. In this study, we have calculated non-elastic cross-sections by using optical model for (n, t) reactions at 14–15 MeV energy. We have investigated the excitation function character and reaction Q-values depending on the asymmetry term effect for the (n, t) reaction cross-sections. We have obtained new coefficients for the (n, t) reaction cross-sections. We have suggested semi-empirical formulas including optical model nonelastic effects by fitting two parameters for the (n, t) reaction cross-sections at 14–15 MeV. We have discussed the odd–even effect and the pairing effect considering binding energy systematic of the nuclear shell model for the new experimental data and new cross-sections formulas (n, t) reactions developed by Tel et al. We have determined a different parameter groups by the classification of nuclei into even–even, even–odd and odd–even for (n, t) reactions cross-sections. The obtained cross-section formulas with new coefficients have been discussed and compared with the available experimental data.     

Contributions of the “Saturn” to Modern Stellarator-Torsatron Research

It has been 40 years since the startup and first plasma from the Saturn stellarator-torsatron. The l = 3 stellarator-torsatron “Saturn” was the first of a series of torsatron-like devices built in Kharkov Institute of Physics and Technology (KIPT) in the 1970s: later appeared “Vint-20”, “U-3”, “U-3M”. The Saturn device was used for comparative investigations on the same device two magnetic configurations in stellarator and torsatron regimes and for experimental examination of their effects on plasma confinement. The thorough measurements of magnetic structure in both regimes demonstrated their high equivalence. Investigations of torsatron without toroidal coils supported the principal possibility to have a spatial divertor configuration, which was later realized in U-3, the first torsatron with a divertor. The results on Saturn, obtained for the first time with a pure torsatron configuration, have opened the prospect for torsatrons to be an alternative to tokamaks in the development of a fusion reactor. After various magnetic configurations were studied, the Saturn device was used for providing investigations of confinement of injected hydrogen plasma and of ECR plasma in different gases. In this paper we summarize the main Saturn results and try to find the bridge between them and the present experiments on existing stellarator-type fusion devices with an aim to see what particular Saturn results were supported by those obtained later.     

The Plasma Jet/Laser Driven Compression of Compact Plasmoids to Fusion Conditions

Magneto-inertial fusion (MIF) is based on both magnetic and inertial confinement. An embedded magnetic field is compressed along with the target plasma to achieve magnetic insulation and fusion condition. Several magnetic systems for plasma confinement may be used for laser-driven (LD) and plasma jet driven (PJ) magnetic flux compression. Estimations show the possibility in principle to realize regimes of PJMIF system with a plasma gain factor Q > 10.     

Exploratory studies of tokamaks as fusion test reactors

Studies have been performed to explore various plasma burn scenarios for a tokamak test reactor which could follow the next generation of large tokamak experiments. Tradeoffs between an ignited burning plasma and a sub-ignited driven plasma are examined in terms of device size and performance as a fusion engineering test facility. It is found that plasma performance levels, measured by ignition margin, amplification factorQ, and fusion power output, increase with device size, more optimistic transport scaling laws, lower magnetic field ripple, and higher. The performance of a generally low stress (B ₀=4 T) reference device, with major radiusR=4.5 m and minor radiusa=1.3 m in a D-shaped (=1.6) plasma has been evaluated over a wide range of operating parameters. In particular, a moderate fusion power output of 300 MW is obtained, the driven plasma havingQ 10, an edge ripple of 1%, and a density ranging between 1.0 and 1.5×10¹⁴ cm^–3. The same device operated at a higher general level of stress (B ₀=5.3 T) is predicted to achieve ignition, but is not required for the mission of an engineering test facility and would entail greater technical risk.     

Utilization of Heavy Metal Molten Salts in the ARIES-RS Fusion Reactor

ARIES-RS is one of the major magnetic fusion energy reactor designs that uses a blanket having vanadium alloy structure cooled by lithium [1, 2]. It is a deuterium–tritium (DT) fusion driven reactor, having a fusion power of 2170 MW [1, 2]. This study presents the neutronic analysis of the ARIES-RS fusion reactor using heavy metal molten salts in which Li₂BeF₄ as the main constituent was mixed with increased mole fractions of heavy metal salt (ThF₄ or UF₄) starting by 2 mol.% up to 12 mol.%. Neutron transport calculations were carried out with the help of the SCALE 4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S ₈–P ₃ approximation. According to the numerical results, tritium self-sufficiency was attained for the coolants, Flibe with 2% UF₄ or ThF₄ and 4% UF₄. In addition, higher energy multiplication values were found for the salt with UF₄ compared to that with ThF₄. Furthermore, significant amount of high quality nuclear fuel was produced to be used in external reactors.     

Determination of Thermomagnetic Current by Using Neutron Released from Deuterium–Tritium Fusion Reactions in Hybrid Reactors

It is proposed to use the neutrons released from a Deuterium–Tritium fusion reaction to drive thermomagnetic currents in a plasma corona surrounding the fusion plasma through the heating of the corona with nuclear reactions by the neutrons released in the fusion reaction because the fusion reaction cross sections are larger for slow neutrons, it is proposed to slow them down in a moderator separated from the hot plasma of the corona, giving the configuration a similarity to a heterogeneous nuclear fission reactor. While in a fission reactor the separation makes possible a growing neutron chain reaction, it here makes possible the autocatalytic amplification of the thermomagnetic currents by an increase of the fusion reaction rate through a rise of the plasma pressure by the magnetic pressure of the thermomagnetic currents. This is expected to substantially increase the product nτ over its Lawson value.     

Combining the “Kinetic Tandem” and the “Kinetic Stabilizer” Concepts

A novel means of MHD stabilizing axially symmetric mirror systems was proposed by Ryutov and demonstrated in the Gas Dynamic Trap at Novosibirsk. It relies on the strongly stabilizing effect exerted by low density plasma on the expanding field lines outside the mirrors. The “Kinetic Stabilizer Tandem Mirror” implements Ryutov’s technique by injecting axially directed ion beams into the expander. The ions, stagnated, and reflected by the converging magnetic field, form the stabilizing plasma. MHD stability code calculations show stabilization at beta values of 40%, with stabilizer beam powers that are small compared to the T-M fusion power output. Implicit in the calculations is the assumption that adequate “communication” exists between the plug plasmas and the stabilizer plasmas. This paper examines one means for enhancing communication: Utilize the stabilizer plasma to create a potential peak that, together with the plug potential, forms a potential well that traps and contains a “bridging” plasma.     

Requirements for D-D tandem mirror reactors

Requirements for D-D barrier tandem mirror reactors are calculated from an equilibrium power balance model. To obtain adequate plasmaQ and reasonable power density, axisymmetric configurations are required to decrease barrier length and radial transport and to increase central cell beta. We find that for a reactor producing 900 MW net electric power from aQ=6.5 plasma, a central cell length of 225 m, maximumB of 15 T, and neutral beam injection energy of 700 keV are necessary. In addition to high central cell beta (70%), high barrier beta (40%) is needed to allow the ECRH power required to reduce the barrier potential. Using too much barrier ECRH power results in a decrease inQ. Nuclear elastic scattering of fusion products plays an important role in the overall plasma power balance. When nuclear scattering and coulomb scattering are included, the plasmaQ value is increased by more than 40% compared to the case when coulomb scattering alone is considered.     

Compression, Heating and Fusion of Colliding Plasmoids by a Z-theta Driven Plasma Liner

A potentially promising approach to fusion employs a plasma shell to radially compress two colliding plasmoids. The presence of the magnetic field in the target plasma suppresses the thermal transport to the confining shell, thus lowering the imploding power needed to compress the target to fusion conditions. With the momentum flux being delivered by an imploding plasma shell, many of the difficulties encountered in imploding a solid metal liner are eliminated or minimized. The best plasma for the target in this approach is the FRC. It has demonstrated both high β, and robustness in translation and compression that is demanded for the target plasma. A high density compressed plasmoid is formed by a staged axial and radial compression of two colliding/merging FRCs where the energy that is required for the implosion compression and heating of the magnetized target plasmoid is stored in the kinetic energy of the plasmas used to compress it. An experimental apparatus is being constructed for the demonstration of both the target plasmoid formation as well as the compression of the plasmoid by a plasma liner. It is believed that with the confinement properties and the high β nature of the FRC, combined with the unique approach to be taken, that an nτ_E T _i triple product ∼5 × 10¹⁷ m⁻³ s keV can be achieved.     

Confinement & Current Drive Measurements for Steady-State FRCs

Detailed measurements in the TCS Rotating Magnetic Field (RMF) driven FRC device display a highly non-uniform resistivity profile, highly peaked near the separatrix where the ratio of electron drift velocity v _de to ion sound speed v _s is large. The RMF parameters determine the plasma density. The plasma temperatures are governed by power balance, and higher temperatures result in higher diamagnetic currents, mostly inside the magnetic field null, and higher magnetic fields, with surprisingly little increase in absorbed power. The results are well modeled by a ‘Chodura’ type resistivity scaling with electron collision frequency scaling as ν_ch∼ω_pi(1− exp[−v _de/v _s]).     

Studies on the Dynamics of D–T Fusion Reaction Based on the Relativistic Equilibrium Velocity Distribution

The Coulomb barrier is in general much higher than thermal energy. Nuclear fusion reactions occur only among few protons and nuclei (i.e., deuterium and tritium) with higher relative energies than Coulomb barrier. It is the equilibrium velocity distribution of these high-energy protons and nuclei that participates in determining the rate of nuclear fusion reactions. In the circumstance it is inappropriate to use the Maxwellian velocity distribution for calculating the nuclear fusion reaction rate. We use the relativistic equilibrium has a reduction factor with respect to that based on the Maxwellian distribution, which factor depends on the temperature, reduced mass and atomic numbers of the studied nuclear fusion reactions. In this paper, we concluded at energy range 10⁵ (keV) ≤ E ≤ 10⁶ (keV), that is smaller than reduced mass energy of deuterium–tritium, m _r c ², the numerical values of R and R _M are not different from each other very much, but by increasing energy near the region of m _r c ² the difference between them are visible, also by increasing energy for example 9 × 10⁶ (keV) ≤ E ≤ 10 × 10⁶ (keV) the difference is obviously more visible. Therefore, we have to use relativistic equilibrium velocity distribution instead of Maxwellian velocity distribution.     

The Study of (<Emphasis Type="Italic">n</Emphasis>,<Emphasis Type="Italic">d</Emphasis>) Reaction Cross Sections for New Evaluated Semi-Empirical Formula Using Optical Model

Nuclear fusion can be one of the most attractive sources of energy from the viewpoint of safety and minimal environmental impact. The neutron scattering cross sections data have a critical importance on fusion reactor (and in the fusion–fission hybrid) reactors. So, the study of the systematic of (n,d) etc., reaction cross sections is of great importance in the definition of the excitation function character for reaction taking place on various nuclei at energies up to 20 MeV. In this study, non-elastic cross-sections have been calculated by using optical model for (n,d) reactions at 14–15 MeV energy. The excitation function character and reaction Q-values depending on the asymmetry term effect for the (n,d) reaction have been investigated. New coefficients have been obtained and the semi-empirical formulas including optical model non-elastic effects by fitting two parameters for the (n,d) reaction cross-sections have been suggested. The obtained cross-section formulas with new coefficients have been compared with the available experimental data and discussed.     

Neon Soft X-Ray Yield Optimization from PF-SY1 Plasma Focus Device

Based on the consideration of that for operation of the plasma focus in neon, a focus pinch compression temperature of 200–500 eV (2.3 × 10⁶–5 × 10⁶ K) is suitable for good yield of neon soft X-rays (SXR), numerical experiments have been investigated on the plasma focus device PF-SY1 using the latest version Lee model code. The Lee model code is firstly applied to characterize the PF-SY1 Plasma Focus. Keeping the bank parameters and operational voltage unchanged but systematically changing other parameters, numerical experiments were performed finding the optimum Y _sxr was 0.026 J. Thus we expect to increase the neon Y _sxr of PF-SY1 from its present typical operation; without changing the capacitor bank and the electrode configuration merely by changing the operating pressure. The Lee model code was also used to run numerical experiments on PF-SY1 with neon gas for optimizing soft X-ray yield with reducing L ₀, varying z ₀ and ‘a’. From these numerical experiments we expect to increase the neon Y _sxr of PF-SY1 with reducing L ₀, from the present 0.026 J at L ₀ = 1600 nH to maximum value of near 26 J at an achievable L ₀ = 10 nH.     

Initial Exploration of High-Field Pulsed Stellarator Approach to Ignition Experiments

In the framework of fusion energy research based on magnetic confinement, pulsed high-field tokamaks such as Alcator and FTU have made significant scientific contributions, while several others have been designed to reach ignition, but not built yet (IGNITOR, FIRE). Equivalent stellarator concepts, however, have barely been explored. The present study aims at filling this gap by: (1) performing an initial exploration of parameters relevant to ignition and of the difficulties for a high-field stellarator approach, and, (2) proposing a preliminary high-field stellarator concept for physics studies of burning plasmas and, possibly, ignition. To minimize costs, the device is pulsed, adopts resistive coils and has no blankets. Scaling laws are used to estimate the minimum field needed for ignition, fusion power and other plasma parameters. Analytical expressions and finite-element calculations are used to estimate approximate heat loads on the divertors, coil power consumption, and mechanical stresses as functions of the plasma volume, under wide-ranging parameters. Based on these studies, and on assumptions on the enhancement-factor of the energy confinement time and the achievable plasma beta, it is estimated that a stellarator of magnetic field B?~?10 T and 30 m³ plasma volume could approach or reach ignition, without encountering unsurmountable thermal or mechanical difficulties. The preliminary conceptual device is characterised by massive copper coils of variable cross-section, detachable periods, and a lithium wall and divertor.     

Practical Optimization of AECS PF-2 Plasma Focus Device for Argon Soft X-ray Operation

For operation of the plasma focus in argon, a focus pinch compression temperature range of 1.4–5 keV (16.3 × 10⁶–58.14 × 10⁶ K) is found to be suitable for good yield of argon soft X-rays (SXR) Ysxr. This is based on reported temperature measurements of argon plasmas working at regime for X-ray output. Using this temperature window, numerical experiments have been investigated on AECS PF-2 plasma focus device with argon filling gas. The model was applied to characterize the 2.8 kJ plasma focus AECS PF-2. The optimum Ysxr was found to be 0.0035 J. Thus, we expect to increase the argon Ysxr of AECS PF-2, without changing the capacitor bank, merely by changing the electrode configuration and operating pressure. The Lee model code was also used to run numerical experiments on AECS PF-2 with argon gas for optimizing soft X-ray yield with reducing L₀, varying z₀ and ‘a’. From these numerical experiments we expect to increase the argon Ysxr of AECS PF-2 with reducing L₀, from the present computed 0.0035 J at L₀ = 270 nH to maximum value of near 0.082 J, with the corresponding efficiency is about 0.03%, at an achievable L₀ = 10 nH.     

Turbulent heat mixing of a heavy liquid metal flow in the MEGAPIE target geometry—The heated jet experiment

The MEGAPIE target installed at the Paul–Scherrer Institute is an example of a spallation target using eutectic liquid lead–bismuth (Pb⁴⁵Bi⁵⁵) both as coolant and neutron source. An adequate cooling of the target requires a conditioning of the flow, which is realized by a main flow transported in an annular gap downwards, u-turned at a hemispherical shell into a cylindrical riser tube. In order to avoid a stagnation point close to the lowest part of the shell a jet flow is superimposed to the main flow, which is directed towards to the stagnation point and flows tangentially along the shell.The heated jet experiment conducted in the THEADES loop of the KALLA laboratory is nearly 1:1 representation of the lower part of the MEGAPIE target. It is aimed to study the cooling capability of this specific geometry in dependence on the flow rate ratio (Q_main/Q_jet) of the main flow (Q_main) to the jet flow (Q_jet). Here, a heated jet is injected into a cold main flow at MEGAPIE relevant flow rate ratios. The liquid metal experiment is accompanied by a water experiment in almost the same geometry to study the momentum field as well as a three-dimensional turbulent numerical fluid dynamic simulation (CFD). Besides a detailed study of the envisaged nominal operation of the MEGAPIE target with Q_main/Q_jet = 15 deviations from this mode are investigated in the range from 7.5 ≤ Q_main/Q_jet ≤ 20 in order to give an estimate on the safe operational threshold of the target.The experiment shows that, the flow pattern establishing in this specific design and the turbulence intensity distribution essentially depends on the flow rate ratio (Q_main/Q_jet). All Q_main/Q_jet-ratios investigated exhibit an unstable time dependent behavior. The MEGAPIE design is highly sensitive against changes of this ratio.Mainly three completely different flow patterns were identified. A sufficient cooling of the lower target shell, however, is only ensured if Q_main/Q_jet ≤ 12.5. In this case the jet flow covers the whole lower shell. Although for Q_main/Q_jet ≤ 12.5 the flow is more unstable compared to the other patterns most of the fluctuations close to the centerline are in the high frequency range (>1 Hz), so that they will not lead to severe temperature fluctuations in the lower shell material. In this case the thermal mixing occurs on large scales and is excellent.For flow rate ratios Q_main/Q_jet > 12.5 complex flow patterns consisting of several fluid streaks and vortices were identified. Since in these cases the jet flow does not fully cover the lower shell an adequate cooling of the MEGAPIE target cannot be guaranteed and thus temperatures may appear exceeding material acceptable limits.All conducted experiments show a high sensitivity to asymmetries even far upstream. A comparison of the numerical simulation, which assumed a symmetric flow, with the experimental data was due to the experimentally found asymmetry only partially possible.     

Calculation of ρR-parameter and Energy Gain for Aneutronic Fusion in Degenerate P–11 B Plasma

Aneutronic fusion reactions are more safe and clean than the other reactions. One of the important candidate for these reactions is P–¹¹ B. This reaction in characteristic conditions creates degenerate plasma. In a Fermi-degenerate plasma, the electronic stopping of a slow ion is smaller than given by the classical formula, because some transitions between the electron states are forbidden. The bremsstrahlung losses are then smaller, so that the nuclear burning of an aneutronic fuel is more efficient. Practical obstacles in this regime that must be overcome before net energy can be realized include the compression of the fuel to an ultra dense state and the creation of a hot spot. In this paper, ρR parameter (Lawson’s criterion) and energy gain for P–¹¹ B are given.     

Towards assembly completion and preparation of experimental campaigns of Wendelstein 7-X in the perspective of a path to a stellarator fusion power plant

The superconducting stellarator device Wendelstein 7-X, currently under construction, is the key device for the proof of stellarator optimization principles. To establish the optimized stellarator as a serious candidate for a fusion reactor, reactor-relevant dimensionless plasma parameters must be achieved in fully integrated steady-state scenarios. After more than 10 years of construction time, the completion of the device is now approaching rapidly (mid-2014). We discuss the most important lessons learned during the device assembly and first experiences with coming major work packages. Those are (a) assembly of about 2500 large, water-cooled, 3d-shaped in-vessel component elements; (b) assembly of in total 14 superconducting current leads, one pair for each coil type; and (c) assembly of the device periphery including diagnostics and heating systems. In the second part we report on the present status of planning for the first operation phase (5–10 s discharge duration at 8 MW heating power), the completion and hardening of the device for full power steady-state operation, and the second operation phase (up to 30 min discharge duration at 10 MW heating power). It is the ultimate goal of operation phase one to develop credible and robust discharge scenarios for the high-power steady-state operation phase two. Beyond the improved equilibrium, confinement, and stability properties owing to stellarator optimization, this requires density control, impurity control, edge iota control as well as high density microwave heating. Of paramount importance is the operation of the island divertor, which is realized in the first operation phase as an inertially cooled conventional graphite target divertor. It will be replaced later on by the steady-state capable island divertor with its water-cooled carbon fiber reinforced carbon target elements.